ML20062B940

Rev 2 to TR-050, Gpun Response to Generic Ltr 88-01 & NUREG-0313,Rev 2

ML20062B940
Person / Time
Site:	Oyster Creek
Issue date:	09/17/1990
From:	Chen C, Lorenzo R, Slear D GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20062B938	List: ML20062B936 ML20062B940
References
RTR-NUREG-0313, RTR-NUREG-313 GL-88-01, GL-88-1, TR-050, TR-50, NUDOCS 9010260262
Download: ML20062B940 (44)

Text

. . . -

Il 9

, I

i GPUN RESPONSE TO GENERIC LETTER 88-01 AND NUREG 0313, REV. 2 TR - 050 REV. 2 I

AUTHORS -

M14 e4.'

C. CHEN - % & D/ NUCLEAR CHEMICAL / MATERIALS ,

i s

U  % d" R.' LOREN2O - MGRY OCEP APPROVALS D. G. SLEAR[

DIRECTOR, ENGINEERING & DESIGN

_ Q 0 ead i J. J. COL Ef DIRECTOR,ENGINEERp[GPROJECTS ,

M C. t --- '

C. A. MASCARI - DIRECTOR

• QUALITY ASSURANCE i

6 9010260262 901018 "

l- PDR ADOCK 05000210 '..

P PNU W, l

t i

I 013-0012.1 L_

d o= = =.

Nuclear TP.-050 REV. 2 TTTLE CPUN RESPONSE TO CENERIC LETTER 88-01 AND NUREG 0313 - REV. 2 W Cm APPROVAL DATE 2 Provides (1) RWCU inspection plan for welds C. Chen 9-8*10 outside the second containment isolation valve, (2) sample expansion criteria for R.L.LorenzodkV O system safe-ends, RWCU outside the second ?f. W. Laggart M . T .I,- /o D. Covill A f */ 7.b containment isolation valve and system weld * * #"

categories, (3) update of 12R inspections and Dbb S*M repairs, (4) revised 13R inspection plan, and (5) revised 13R piping replacement and stress improvement plan. The content has been revised entirely.

1 l

l in N0036 (03 90) f

. TR ~ 050 s' R3v. 2  !

Page 2 of 43

)

TABLE OP cunannan PAGE

1.0 INTRODUCTION

l 1.1 Purpose 4 1.2 4 Summary of Oyster Creek ICSCC Inspections 4 1.2.1 s 10R and 11R Refueling outages-Inspections / Repairs 4 1.2.2 12R Refueling Outages - Inspections / Repairs 5 1.3 Summary of Proposed 13R Inspection Plan 6 1.4 Summary of Planned ICSCC Mitigating Actions -

6 2.0

SUMMARY

OF SPECIFIC PIPING SYSTEMS 7' 2.1 Recirculation System 7 2.1.1 System Description, Materials and operations.

7 2.1.2 Background / History 7

2.1.3 12R Scope of Work 8

2.1.4 Future Improvement / Inspection Program 9

2.1.4.1 Future Improvements 9 2.1.4.2 13R and Future Inspection Program 9 2.2 Core Spray System 9

2.2.1 System Description, Materials and operations 2.2.2 9 Background / History 10 2.2.3 12R Scope of Work 2.2.4 10 Future Improvement / Inspection Program 10 2.3 Shutdown Cooling System 11 2.3.1 System Description, Materials and operations 11 2.3.2 Background / History 11 2.3.3 12R Scope of Work 2.3.4 -11 i Future Improvement / Inspection Program 11 2.3.4.1 Future Improvements 2.3.4.2 '11 13R and Future Inspection Program 11 2.4 Reactor Water Clean-Up (RWCU) Syst&m 11 2.4.1 System Description, Materials and Operations 2.4.2 Background / History 11 12 2.4.3 12R Scope of Work 12 2.4.4 Future Improvement / Inspection Program 12 2.4.4.1 Future Improvements 12 2.4.4.2 13R and Future Inspection Program 12 2.5 Isolation Condenser System 12 2.5.1 System Description, Materials and operation 2.5.2 Inside Containment 12 13 2.5.2.1 Background / History 13 2.5.2.2 12R Scope of Work 13

! 2.5.2.3  ;

Future Improvement / Inspection Program 13 2.5.2.3.1 13R and Future Improvements 13 2.5.2.3.2 Future Inspection Program 13 013-0012.2 -

1 l

,w _ _ m 4+e- - , a & a , - s- - + - ,,, a

. 1

. .TR - 050 '

s- R v. 2 Page 3 of 43 L

thGK 2.5.3 outside containment 2.5.3.1 Sackground/ History 13 2.5.3.2 12R scope of Work 13 2.5.3.3 Future Improvement / Inspection Program 14 ;

14 2.6 Closure Head Piping Wolds 1 2.6.1 Description, Materials and operation 14 2.6.2 12R scope of Work 14 ;

2.6.3 Future Improvement / Inspection Program 15 .

15 3.0 WATER CERNISTRY COWTROL AT OYSTER CREEK 15 3.1 GPUN Actions Taken 3.2 15 Water Chemistry Effect Data from other Utilities 16 4.0 GPUN TECENICAL CLARIFICATIONS TO OL 88-01 17 4.1 10R and 11R Inspections 17 4.2 Post stress Improvement Inspection 17 4.3 Inspection of Cast Materials 4.4 19 4.5 Implementation of Hydrogen Water Chemistry' 19 RWCU Welds outboard of the Second Insolation valve 20 5.0 NUREG 0313, REY. 2 t 5.1 NUREG 0313 $ cope 21 5.2 GPUN Proposed Program 21 5.3 sample Expansion 21 22 6.0 DTEER GENERIC LETTER 88-01 RESPONSES REQUIRED 23

'0.

SUMMARY

s 26

8.0 REFERENCES

26 9.0 FIGURES 27 10.0 TABLES 32

+

+

013-0012.3

__ ~

I

9 0

,. TR - 050 Rev. 2

. Page 4 of 43 i

1.0 INTRODUCTION

2 1.1 Purpose ,

CPU Nuclear (GPUN) is required to perfor !**;ections of reactor coolant systems (RCS) piping to dete'a intergranular stress ,

corrosion cracking (IGSCC). For the 12R rtfueling outage, GPUN planned to meet the intent of Genrric Letter 84-11[1], which was ,

the guideline for such inspectiois until recently. In January 1988, the Nuclear Regulatory conmission (NRC) issued Nuclear Regulatory Guide (NUREG) 0313 R'iv. 2[2), and this was suppl 6mented by Generic Letter (GL) 88-01[3] In accordance with the generic letter, GPUN's response was submitted on August 12, 1988, out-lining future inspection program plans including the 12R refueling outage. In response to our 12R plans the NRC issued a letter [9]

which took exceptions to our plan. Briefly, these exceptions included: (1) the exclusion of the RWCU piping volds outside of the second containment isolation valves; (2) taking credit for 10R inspections unless the examiners who took the requalification-exams had passed on the first attempt; (3) the sample expansion criteria (unless further tt:hnical justification is provided) for the recirculation system safe-ends, isolation condenser piping outside the second containment isolation valves and the RWCU piping outside the second containment isolation valyse; and (4) the number of category G welds remaining after 13R. +

Subsequent to the receipt of the NRC letter (9), GPUN initiated two telecons [10, 11) to clarify the NRC requirements for the 12R and 13R inspections. Based on these telecons, GPUN committed to revise its inspection plan [13), address the remainder of the category "G" welds by the end of 13R [13), and address the RWCU l

l piping welds outside the second containment' isolation valve six  ;

months after restart from the 12R outage [14). On April 17, 1990, '

NRC issued a safecy evaluation of the above mentioned GPUN's {

submittal and requested GPUN to submit revised inspection plans at l 1 east four months prior to the 13R refueling outage [18). '

This document will delineate GPUN's revised IGSCC inspection '

l program which includes: (a) an update of the 12R outage IGSCC inspections and repairs; (b) the proposed IGSCC inspection plan for 13R and future outages; (c) the planned mitigating actions to minimite the possibility of IGSCC.

1.2 . Summary of Ovster Creek IGSCC Inspections 1.2.1 10R and 11R Refuelina Outanes - Inspections /Renalra In 1983 (10R), 31 Recirculation System welds were inspected to IEB 82-03[4), including three welds that l

were fluorescent dye penetrant inspected'on the ID and dispositioned as geometry. No indications of IGSCC were detected [5).

013-0012.4

' l 1 . 1 l

,- TR - 050 Rev. 2

. Page 5 of 43 In 1984 (10R), a leak was detected from an Isolation 5 condenser condensate return line weld during a pressure test of the condenser tubes. This leak.resulted in the ,

ins;*ction of over 150 welds in the Isolation Condenser System (ICS) and the RWCU System. All the inspectable steam and condensate Isolation Condenser welds outside the drywell (127) were inspected. Twenty-seven welds in the ICS piping outside the drywell were found to contain indications of ICSCC. Nine were replaced thru spool piece change out and eighteen were repaired with full structural weld overlays (6). Three welds were destruc-tively examined, and the failure mechanism was concluded e to be intergranular stress corrosion cracking.

In the 11R refueling outage, inspections of Reactor Coolant System (RCS) welds were performed following the guidelines of Generic Letter 84-11. One hundred and I sixty-nine (169) butt welds which included all eighteen i (18) ICS weld overlays deposited in 10R were inspected. i Three welds in the C-loop of the Recirculation System and one weld in the Isolation Condenser System (ICS) steam line outside the Drywell had indications of ICSCC. The indication in the ICS weld was determined to have been a 1 misinterpreted indication during IOR and not a "new" crack (7). The initial inspection sample as well as the  !

increased sample inspection welds were chosen based upon -

1 the previous inspections and the difficulties associated

{ with determining their final disposition as non-ICSCC. i One Recirculation system weld was evaluated against the criteria of the then-draft Rev. 2 of NUREG 0313, and accepted as stress improved (SI'd); the other two Recirculation system welds and the ICS weld were repaired with full structural weld overlays.

l 1.2.2 11R Refuelino Outaaes - Inspections /Renairs During the 12R refueling outage, 156 butt welds within the scope of Generic Letter 88-01 in the Recirculation, Shut-i down Cooling, Reactor Water Cleanup, Core Spray, and Isolation Condenser systems, closure head piping, and six structural weld overlays (two in the Recirculation system and four in the Isolation Condenser system outsido'the ,

drywell) were ultrasonically examined for IGSCC.- Six butt welds contained UT indications with the chkracteristics of IO SCC.' Five (two Recirculation system and three in the- i Isolation Condenser system outside the drywell) required weld overlay repair;~one' Recirculation system weld was analysed and found to be acceptable for continued' opera-

tion without repair. The six. inspected overlays contained i no indications of ICSCC in either the overlays or the 013-0012.5

. _ E. . - . -

[ TR - 050  ;

Rev. 2

, Page 6 of 43 l

outer 25% of the original pipe wall. Additionally, two welds (one 8" Core Spray, one 2" Reactor Head Cooling '

[RHC)) were found to be leaking during the hydro test.

The core spray weld was overlay repaired; the head weld was cut off and re-welded. The RHC defect was destructively evaluated and found to be IGSCC.

1.3 Summarv of Proposed 13R Insometion Pita During 13R refueling outage, eighty-four (84) pipe welds plus eight (8) safe-end welds will be inspected. Additionally, ninety-one (91) replacement welds (Category "A") will be UT  ;

baseline inspected. These welds are listed in Table 2 through  !

Table 14.

1.4 Summary of Planned ICSCC Miticatina Actions 1.4.1 Implement Hydrogen Water Chemistry (NWC) during cycle 12  ;

(1990). As a result of implementing HWC, a reduced inspection frequency may be requested in the future, ,

based on industry experience and BWR Owners Group's recommendations (19).

l l 1.4.2 Stress Improve (SI) all accessible /inspectrble welds l' inside the drywell (except Reactor Water Clean-Up System) by the end of the 14R refueling outage (1992).

Forty-seven (47) welds will be stress improved during 13R and the romaining eight (0) recirculation safe-end welds will be stress improved in 14R.

1.4.3 Replace the following pipes with IGSCC resistant material during the 13R refueling outage (1991).

a. all Isolation Condenser large bore piping outside the drywell (from the drywell penetrations to the iso-lation condensers) that has been very susceptible to IGsCC. This will reduce the number of welds requiring l inspection by sixty-seven (67). In addition, the six outside containment isolation valves will be replaced. i

b. all piping within the four (4) isolation condenser drywell penetrations and the two (2) Reactor Water 4 Clean-up system drywell penetrations which contain welds that are not inspectable. -This will eliminate thirteen-(13) presently uninspectable welds,

c. all Reactor Vessel Closure Head pipes with diameter larger than 4".

I 1.4.4 Corrosion Resistant Clad (CRC) the three (3) reactor vessel closure head nostle-to-flange welds.

013-0012.6

?

~ . . -- . . -- - - - . - - - - _

. l

.* TR - 050 l Rev. 2 i Page 7 of 43 I 1

2.0 StnetARY or SPSCIF',G_U 7ING SYSTEMS 2.1 Recirculat;&p System 241.1 system Description. Materials and coerations

\

Dyster Creek's' Recirculation System consists of 26" OD .

piping fabricated with Type 316 heavy, wall stainless. '

seeel. The entire system experiences large flow rates of-water at operating temperature and pressure while tho' 't reactor is in operation. The system consists of five i loops with piping of uniform dimensions. Unlike later vintage boiling water reactors (BWR's), each loop is segregated from the others and takes suction from the; '

reactor vessel annulus and discharges to the lower vessel area containing the diffuser. t 2.1.2 Backaround/ History IGSCC in this system was described' earlier in Section 1.2.1.

During the 11R refueling outage of 1986, 64 of the> system's 89 welds were stress improved (SI'd)'with the induction heating stress improvement (IHSI) method. All.64 welds were inspected following. stress improvement. Among-the s 64 welds inspected, three (3) were found to have indications-of IGSCC. Two of the ICSCC affected welds were overlay repaired and one remained in the use-as-is condition.

Uninspected and non-stress improved welds include 20=

safe-ends and b uninspectable castir.g-to-casting welds..

The Recirculation System safe-ends /were overlayed on both the ID and OD before operation began in 1969[8).- This was performed because cracking was' detected in several Type 316 components that were subjected to the final vessel: heat treatment resulting in the sensitizing of these components.

The overlays (both ID and OD) were " low: carbon,;high- '

ferrite" (as stated in the repair: specifications) Type 308L weld metal. However, since the safe-end to nozzle shop weld l was performed with Inconel 182 weld metal, a portion of the (

ID is covered with Inconal 182 (see Figure 1). i Since we cannot locate the weld material chemistry test reports nor the inspection reports of the recorded, as-deposited ferrite content, we consider it prudent to stress improve the safe-end to piping system welds.; And, since the ID of the nozzle to safe-end welds were overlayed with-Inconel 182,.we consider it prudent to stress improve the ,

nozzle to safe-ends welds. Approximately 23 man-rom of' exposure per safe-end will-be' required to perform?these tasks.

i 013-0012.7

.. . s l: ,

• 1 s

TR - 050 '

i .* R0v. 2 Page 8 of 43 L 2.1.3 12R - Seone of work  !

During 12R, <

10% of,the 61 previously streep' improved (SI'd) welds plus the three welds containing ICSCC were~ inspected.

The initial sample of 6 category "C" welds inspected in'12R-had 2 with indications of ICSCC (No-D-11 and NO-D-21). As  ?

a result of this finding, an' additional 6 welds required inspection.

In this additional sample, one weld (NG-D-18) a had indications of IOSCC.- The methodology utilized to determine the initial welds as well as the first sample expansion was based upon previous inspections and'the difficulties position.

associated with determining their-final dis-Therefore, the remaining 49 category "C" welds l were inspected during 12R. No'further IOSCC indications ,

l were detected. All 61 Category C welds were reinspected in 12R.

Also-during 12R, both."C" loop recirculation' safe-ends:were I stress improved and inspected. The effort required to

[

perform this included machining the OD c16dding to' provide a surface finish and contour. adequate for. performing UT for IOSCC.

These weldsLwere then poet' process UT examined.

The inspection area included the' nozzle to safe-end weld (the safe-end side of that weld,-and'the nossle side of that, weld for a distance of IT into the nottle-from-the weld centerline) and the safe-end-to-pipe' weld.--_These ,

welds contained no' indications of IGSCC. (

The technical basis for selecting these'two safe-ends was >

that-during the 11R' outage, CPUN stress improved 64 Re-circulation System welds and-inspected all'of'them'atter.

SI. L

.The only loop to contain' welds'with indications =of

.IGSCC were three in the C-loop.' During'the' safe-end overlaytecladding effort'before operation, described above, accese n

.the vessel was provided'by removing theJelbow at the top of the C-loop vessel inlet riser. 'one-of the replacement welds (No-C-23) was-one of the three welds- j

.found to contain indications of:IGSCC during 11R. -We  !

consider that.these circumstances-previded sufficient- ,

concern first.

to warrant stress improving;these two safe-ends- ,

Treating,these two safe-ends during 12R has provided.much l

needed, useful.information for stress improving, machining, and inspecting the remaining eight. safe-ends in future-outages. Knowledge of the radiological; environment and 1essens, learned during-the 12R outage will' enable us to- q more efficiently perform this work in the future.' Wo. j expect that future time and' exposure savings will ius substantial. j

.Since no^ indications of'IGSCC were found

• in the C-loop safe-ends'in'12R, we. consider that these l

'013-0012.8  !

al -

un

__ L'I'

,, 'TR~- 050 <

Rev.' 2 Page 9 of 43 l results coupled with the partial' implementation of HWC in cycle 12 provides adequate assurance that cracks will not "

initiate / grow in the other eight safe-ends.

During 12R, 3 D-Loop welds were'dispositioned'a's containing indications of ICSCC'(NG-D-11, 18,.and 21). A. plug was ,

removed from NC-D-11 during.12R for destructive evaluation.

l Results of this analysis revealed that the crack tip was blunted and probably existed' prior to IHS! in 11R-(15).

Welds NG-D-11 and NG-D-21 were overlay repaired and NG-D-18 was left as stress improved.

2.1.* Future Imorovement/Insneetton Procram 2.1.4.1 Future Imorovements - The planned mitigating  ;

actions for the recirculation system welds ,

include' implementing HWC following 12R,. stress' .!

improvement, and post) process inspection of the safe-ends in A,~B, D, and E loops of the i recirculation system. Four safe-ends will be +

stress-improved and post process inspected during i 13R. The remaining four safe-ends will be stress )

l improved and' post-process inspected during 14R. . i 2.1.4.2 13R and Future Inspection Procram - The welds of recirculation system-will-be inspected in accordance with GL88-01 and, Tables 2 and 5 of this report. As discussed;incref.1(19),~only eight of the sixteen uninspected safe-end category G welds!will be stress improved and inspected during :13R.E The remaining-8 welds will be stress-improved.and inspected during 14R.

2.2 Core Sorav System 2.2.1 System Descriotion. Materials and ooerations l

l The Core Spray System is designed only to be in use during l

a loss-of-coolant accident (LOCA) involving a loss of'  !

reactor water inventory. Generally,^the system has a.

temperature environment of less than 200'F. Because of  !

L thermal mixing, some piping and welds close to the reactor.

l vessel exceed 200'Fi therefore, boundaries of-IGSCC susceptibility are limited. .The-system external to the-reactor vessel' consists of 6 incht and 8 inch' diameter Type-316 stainless'ateel pipe. Having two redundant' loops, the system enters the reactor vessel through two safe-ends ,

attached to two separate nozzles.' The system has a total l

of 27-potentially ICSCC susceptible welds'in the scope of-NUREG 0313 Rev.2.

013-0012.9 .

l 4

]

.- i

, TR - 050  !

Rev. 2 I

. Page 10 of 43 l l-l 1

l 2.2.2 Backaround/ History --

During the 10R outage, 3 Core Spray. System welds were inspected for IGSCC. In 11R, 16 welds were inspected, of which 2 had been inspected in 10R. No indications of IGSCC were detected.

( 2.2.3 12R Scope of Work I

l During the 12R refueling outage, the system's safe-end welds and 19 of the 21 IGSCC susceptible piping. butt _ welds

, were stress improved. Initially,.11 core spray welds were

(

inspected.L This number included the 6 safe-end welds associated with the systems 2 safe-ends. Within this-initial group of welds, there were no indications of l IGSCC. Not all stress improved-welds received a post- _  ;

= process inspection,'but it was assessed that a sufficient sample were post-process inspectedL(See section 4.'2_for l Technical Basis). However, all safe-end welds were:in-l _

spected after stress improvement. Subsequently, during:a pre-operational hydrostatic pressure test, a leak _was detected from core spray weld NZ-3-38. ' Subsequent exami-nations identified the presence of indications of.IGSCC contained in this weld. This weld, NZ-3-38, was one of-the 19 welds which were; stress improved during 12R. It:

was determined that NZ-3-38 was examined during 11R, but did not receive a post stress improvement inspection, during 12R.- As-a result of~this event, an expanded-sample group of eight-(8) additional welds (NZ-3-39, 40, 42, 81, 84, 88, 89, & 90) were examined. Within-thisigroup seven-(7) were stress ~ improved during 12R.- Within this population, two (2) welds (NE-3-40, 42),were pre-process' inspected in 11R,-however none were post-process inspected during 12R until this event. There were no indications of- '

IGSCC detected within the. expanded: population. <

NZ-3-38 was weld overlay repaired and returned to service.:

2.2.4 Future Imorovement/ Inspection Procrami 25 of 27 welds within the scope of NUREG 0313 wore stress.

s improved during 12R.~ The 2 remaining welds will be SI'd and inspected in 13R. Stress improved but not previously inspected (11R or 12R) welds will be inspected during the 13R outage., The system's safe-ends and pipe welds will be-inspected in accordance with GL88-01 and Tables-3 and 7 of 1 this report.

t 013-0012.10 d

1

)

TR - 050-

.* R;v. 2 Page 11 of 43-

=

l J

2.3 Shutdown Coolina System i

2.3.1 System Descriotion, Materials ~and coerations- >

The Shutdown cooling system is utilized' intermittently, only during plant shutdowns where cooldown to below 212*F is required. However, initiation may be at temperatures as elevated as 350'F for very short periods of time. The system is composed of 14 inch diameter Type 316 stainless steel, schedule 80 pipe (14 welds total), and maken a  :

transition to carbon steel inside the drywell before the 2nd containment Isolation Valve (CIV).

2.3.2 Backaround/Histogy This system has shown no indications of IGscc thus far..

Two of the welds in shutdown cooling were inspected in 10R. During 11R, 6 welds were examined, of which 2 were' examined in 10R. The system contains a total of 14 susceptible welds.

2.3.3 12R Scope of Work Three welds in shutdown cooling were inspected during 12R~

with no indications of ICscc. In addition, NWC, which will

+

I-be implemented: following' the: 12R refueling outage,- will be a significant IGscc mitigator for 9 of'the 14 welds (see section 4.4 for Technical Basis).

2.3.4 Future Imorovement/ Inspection Procrams 2.3.4.1 Future Imorovements - The 14 welds will be stress improved ~and post process inspectedEduring the.

13R refueling outage.

2.3.4.2 13R and Future Inspection Procrams - All the welds will be? inspected.in accordance with

• GL88-01 and Table.6 of this report.-

~

2.4 Reactor Water clean-Uo-(RWCU) ~ System' 2.4.1 System Descriotion, Materials and Operations The RNCU System is. operating at virtually all.-times'during power operation. It consists of 6 inch diameter Type'316-stainless steel pipe.-.Up to.the inlet of the first non-regenerative heat exchanger and from tho' outlet (return) of the third regenerative. heat exchanger, reactor coolant is above 200'F.

i 1

013-0012.11

.- - , - a

l

.TR - 050-

.. R;v. 2 Page 12 of 43 2.4.2 Backcround/ History During the 10R refueling outage, five (5) welds were inspected in the RWCU system. A total of 10 welds inside-containnent were inspected during 11R of which 2 had been -

inspected in the 10R refueling outage, No IGSCC has been detected in the system to date.-

2.4.3 12R Scope of Work 10 RWCU System welds inside containment.were inspected ~

during 12R with no' indications of IGsCC.- In addition, HWC, which will be implemented following the 12R refueling- l outage, will be a significant IG8CC mitigator.

(

2.4.4 Future Imorovement/Insometion Procram 2.4.4.1 Future Imorovements'- Since RWCU is operating during plant power operation, the system will' derive full benefit from HWC. In addition, the piping that is inside. containment penetrations (unaccessible for UT inspection) will'be' replaced with IGSCC resistant ^ material in 13R.. The new piping will contain-no butt. welds inside the penetration.

2.4.4.2 13R and Future Inspection'Procram : All the RWCU welds located inside:the drywell+will be-inspected in accordance with1GL88-9 and' Table 8.

10% of the RWCU welds outboard of the second isolation valves will be inspected during 13R (18, 19). 1 2.5 Isolation Condenser System 1

2.5.1 Svatem Descriotion. Materials and coeration i The, Isolation Condenser System (ICS) is a standby,;high pressure system for removal of fission product heat from the reactor vessel following a reactor trip.and for-iso- L lation of the reactor from the main condenser. The system prevents overheating of'the reactor fuel;Jcontrols the re-actor pressure rise, and limits the loss'of reactoricoolant' ,

l through the relief valves.. During normal = power operation, l l

I the system is fully pressurized; however,' flow is prevented i during the standby mode by one closed condensate returni ' l line isolation valve _per loop. Tho' system. consists of 10",.

12" and 16" Type 316 Stainless Steel on the' steam' side,,and 8" and 10" Type 316 stainless steelion the condensate side.

The ICS has experiencedznumerous' initiations in,the ye'ars ,

since initial start up of oyster Creek.

013-0012.12

-~ . . .'.

. TR'- OSO

.* R;v. 2 j Page 13 of 43 I

i

-J 2.5.2 Inside containment- 4 2.5.2.1 naekoround /Historv_ ,No IGSCC has been detected in this portion of the ICS. During 10R, 18 welds -]

had been inspected.. In 11R, 12 welds were in-- ]

spected of which 6 had been examined during.10R.~ .l 2.5.2.2 12R Scope of Work - During 12R, a total-of 14 1 q

welds which' include the 2 ICS safe-ends (a. total of 4 welds) and-10' piping welds were inspected ',

with no indicatir ns of IGsCC. Stress improvement e

was performed o9 the 2 ICs safe-ends (4 welds)-

and on 9 additional steam side welds. Not all.  ;

stress-improvvd, welds-received a post-process-e

j inspection- bvt a sufficient sample were post- '

-process insp9cted (see Section 4.2 for Technical Basis). However,.all-safe-end welds were in- ,

spected after stress = improvement. In addition, f{

HWC, which~was implemented following the 12RL '

refueling outage, will be a significant ICSCC' mitigator.for the condensate piping =up to the second (normally closed) valve. '.

2.5.2.3 Future Inorovement /Insometion' Procram 2.5.2.3.1 Future Imorovements - In 13R, all of the welds not yet stress improved will be treated.- Also during 13R, the piping within the'four-(4) ICS containment penetrations will be replaced'with resistant material with= i no butt: welds,inside the penetrations.

2.5.2.3.2 13R and-Future'Insoection Procram --

All the welds will be inspected in  :

accordance with GL88-01 and Table 9, 2.5.3 Outside containment' ,

2.5.3.1 Backaround/ History -

In 10R,.127 ICS welds outside of containment were examined. .In this case, a large sample was inspected because of detecting.a through-wall leak in an 8" dia.-  !

, condensate returntline weld.. As a result of this-

, effort, 18 welds were overlay repaired and 9 were ]

replaced through spool' piece changeout. During l  ?

11R, 58 previously inspected welds were examined.

Of.these, 18nwere the'10R overlayed welds. One-I 013-0012.13 l I

\

1 at -1

~

s TR - 050 Rev. 2 -

Page 14 of 43

]

l- . -

additional. weld was found to contain indications l

L of IGSCC, and as a result, was'overlayed in 11R. 1 This weld was diagnosed during 10R as having a root geometry indication. Seventeen (17) more welds that were installed as a result .of' apool j l

piece repairs in 10R were not included in theellR

-sampling base.-

l 2.5.3.2 12R Seone of Work - During 12R, GPUN inspected 37 welds in ICS outside containment.; :Within this sample, there are 4 overlayed welds,(one was overlayed in 11R), 23. welds which were'not ,

inspected during 11R, and 10_ welds inspected during 11R.' This inspection resulted in 3 welds reqoiring weld' overlay repair.- Two were not-inspected during 11R and one-was inspected during 11R. The sample expansion,.as a result of the above indications, was,in accordance with NUREG 0313,-Rev.2 for the ICS-Piping,outside:the-Drywell.

2.5.3.3 13R and-Future Imorovement/Insnection Procram '

All the piping outside the drywellLand;inside I the penetrations will kan replaced in 13R with ICSCC resistant material. Thesel welds'will be-inspected 'in accordance with. 0L88-01.and Table 12 of this report. All new welds outside containment will also be stress improved (where i possible)-in: order'to'further,reducetthe potential for :IosCC. . t' i

2.6 Closure Head Picina Welds 2.6.1 Descriotion, Materials, and operations-Therearethree-alloysteel"nossleswithIType)316 stainless steel weld neck flanges welded to them on.the reactor-vessel closure head. -The origina1Lflanges were subjected to the closure head's final post weld. heat' treatment. I Thereby, the flanges were furnace sensitised. 'All three flanges were-replaced as part of the' pre-operation repair effort described earlier.. The nossle weld preparation was buttered with Inconel 182, and,the replacement, flanges butt welded with Inconel 82; one 6"~nossie ista: spare to which a blind-flange is. bolted. .The:other J" nozzletis the-inlet for the head spray line. Thel4" nozzle is'for. vent piping. .

There are five (5) other butt welds connected'toLthe above nozzles which are susceptible to IGSCC. Therefore, the closure head contains-eight (8) welds within.the boundary 1 l of NUREG 0313 Rev. 2.

013-0012.14

1

,. TR - 050- .,

.Rev. 2 j Page 15 of 43: l I

l 2.6.2 12R' Score of Work +1 i

l During the 12R refueling outage 2 of!the closure head nozzle-to-flange welds were inspected with no indications l of IGSCC. However, subsequent to'these. examinations a leak-was observed adjacent to a 2 inch butt weld within the head spray nozzle 6 x 2 inch reducer during the hydro test. .As i a result of-this observation a ring. sample was removed con- ,

taining the defect and destructively analyzed. The results  ;

of the analysis characterized the crack to be IGScc (16).'

~

An expanded examination was= conducted where two four-inch. ,

and one two-inch butt. welds within the vent' piping were I ultrasonically examined >to determine:the presence of IGScc.'

The results of these. examinations showed no indications of IGSCC were present.- The head spray nozzle weld was.

repaired and returned to service. i j 2.6.3 13R and Future Imorovement/ Inspection Procram We plan to improve all welds to category A by the end of the 13R outage. The three nozzle-to-flange. welds will be j corrosion resistant clad (CRC) with resistant material. The ,

remaining piping will be replaced with materials resistant $

to IGScc. Applicable. welds will;be. inspected in.accordance ,

with Generic Letter 88-01: and : Table 10..  !

3.0 WATER CHEMISTRY CONTROL AT OYSTER CREEE 3.1 GPUN Actions Taken i

4 The following actions have been.or will be taken to mitigate the initiation and propagation of'IGSCC at Oyster Creek l Implemented EPRI water chemistry guidelines (1984). t commence hydrogen water chemistry (HWC) during cycle 12 (1990).

Established a new chemistry laboratory with. state-of-the-art'-

equipment for analyses (1985).

Plugged a large. number of. leaking condenser tubes during 10R.

During cycle'11,.no. tube leaks were evident.o

a.  ;

Replaced resin in the reactor water cleanup'demineralizer at the onset of silica: leakage.

Perform air in-leakage surveys.

• Replace resin in condensate. polishers before onset of

significant ionic leakage (1985).

[

013-0012.15

. .m .

TR - 050 Rev.=2

. Page 16 of 43 our efforts to improve water. chemistry have.resulted in substantial improvements over the last two operating cycles. -Reactor Water.

conductivity during cycle 10, Cycle 11, and cycle 12 to date has been about 0.1 uS/cm. We consider that this improved conductivity has substantially reduced the potential for new crack initiation and.has slowed the growth of existing, undetected cracks, if any.

~

The addition of HWC in Cycle 12 will further reduce the potential for new cracks..

Pipe tests sponsored by EPRI & GE indicated a factor of improvement of 20 with respect to crack initiation following the introduction of appropriate additions of hydrogen.

We also consider that the lack of ICSCC in' the shroud head bolts (creviced alloy 600) is an indication of' adequate water' chemistry performance in the past. In other BWR's, many-of these bolts, identical f in design to oyster Creek's, were found to contain crevice-induced-IGSCC, including some that were 1004' cracked. The cause of this cracking has been attributed to the presence of a crevice beneath the collar, and water conductivity. In.the-11R outage,-all-36' bolts were UT inspected at Oyster Creek. No indications were detected.. Again, the improvements in water chemistry: control noted earlier'are. expected to provide significant improvement in reducing the potential 1for new crack, development.

3.2 Water chemistry Effect Data From Other Utilities' In-plant water chemistry studies performed at.Dresden-2 and Peach Bottom-3 have shown the significance ofLimproving normal water j chemistry control (NWC), specificallyffor conductivity. This work,.

which is-partially funded by'EPRI, istbeing! performed to show the.

improvement on materials performance'when good chemistry practices i

are followed. The Dresden-2 data has been obtained:while the plant was injecting hydrogen (HWC) ~ and the:PB-3? data was.obtained'under HWC conditions. Figures 2 and 3 show the average crack growth ra es for sensitized Type 304 specimens loaded to about 27 Kai -

in .

Figure 2 data for PB-3 clearly shows!the improvement'in crack ~ .  ;

growth when water conductivity is; reduced.; .There is,a factor of 4 1 improvement (from 96 to 24 mils /yr.)lwhengthe averaos conductivity-3 is reducad from 0.5 to 0.2 uS/cm.- The'FigureL3'PB-3 data shows the 3 i; impact of two resin intrusions on. growth while the normal chemistry~

t! conductivity had been maintained'at"about 0.2 us/cm. 'Even with the L two resin intrusions within.a short period:ofotime apart (about 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />), the average crack growth rate increased by only a factor of 2 (from 22.5 mils /yr. to 44 mils /yr.) ;We would expect that'barring a further upsets, the crack growth rate would return to its rate before'the intrusions. These data were.obtained'from CAV specimens in autoclaves connected to the plant's,RCS.. d

.i j

q j

013-0012.17 t

li ' ,h

' 11

. - e

1 1

TR 050 l

• ^

rov. 2

, Page 17 of 43 Similar results have been obtained in_the laboratory. The crack l' growth rate for sensitized Type 304 at about 0.1 uS/cm conductivity was about 29 mpy while at about 0.45 uS/cm, the rate was about 240 mpy (Figure 4). These growth' rates compare very favorably with the NRC-calculated rate of.390 MPY (Appendix A.2 of-l NUREG 0313 Rev. 2). Figure 2's Dresden-2 data shows the' impact of

( implementing HWC with low conductivity (about 0.1 uS/Cm). . The  ;

I Figure 4 data shows similar results in laboratory conditions. We I consider that the low average conductivity obtained at Oyster .

Creek over the last few cycles has significantly contributed to'a

• reduced propensity for developing _new cracks and aislow crack'

growth rate for potential existing cracks; therefore, even if cracks have initiated, growth will besolow and does not represent l a safety concern. The implementation of HWC during Cycle 12 (1990), will further improve the condition of affected stainless steel piping.  ;

4.0 CPUN TECHNICAL CLARIFICATIONS TO CL88-01 4.1 10R and 11R Inspections The NRC staff did not concur with GPUN's position for taking credit for the 10R outage examinations, because these examinations'were-performed prior to September 1985. The staff did agree, however,

( to consider those 10R inspections if GPUN could demonstrate that the examiners had passed the:requalification tests:on the first; l attempt. GPUN has reviewed the records, and cannot demonstrate I

• that the 10R examiners took the requalification exam.. Based on this fact, CPUN will not take credit for the 10R-inspections.

However, we note that the 10R inspections did,'in fact, detect 1 IGSCC in 18 welds and " suspected" IGSCC in'9.

Of these,> cracking was verified; destructively in 3, non-destructively by ID PT-examination of replaced welds in 5, and one suspect;IGSCC; call was destructively verified as not being IGSCC.- Therefore, while these examinations were performed by examiners not considered. qualified,.

it-is clear that they were effective in detection =offIGSCC.

4.2 Post Stress Imorovement Inspections 1

We consider that for certain sizes of piping. performing stress-  ;)

improvement.(SI) without 100s immediate post-SI. inspection is a prudent technical approach to mitigating'IGSCC and that performing a 100% reinspection over the following'two'outagesjis not; warranted. .I The concern with not performing immediate inspection is.that, while

'SI places the inner half of the weld in compression, it also places .

the outer half in tension. For large-diameter piping:(> 12-inch), -

the residual stress near mid-wall is largely compressivei(about - 15 y Kal). This is the reason that most-IGSCC in large diameter piping l l

013-0012.18

'l ,

c., ~ i

• • TR - 050 i R3v. 2~

, Page 18 of 43 appears to have been arrested near mid-wall. Changing the residual f stress pattern such that the outer half is subjected to tensile stresses could, in fact, result in continued through-wall. growth of ,

a previously arrested crack. Although we are.not-aware of any cases where this has happened, we consider this to be a valid technical concern. Therefore, for all 14-inch shutdown Cooling welds which are stress improved, a post process inspection will be perform sd.

However, for smaller diameter piping (<12-inch), calculatfsne show that a 10% through-wall-(TW) crack will propagate to 80% TW iiithin one operating cycle.- The maximum crack depth allowed by section XI is 60% of pipe wall thickness. .This is a result of the prveence of a linear TW residual stress that is tensile on the ID and compressive on the OD. Once the crack reaches mid-wall Lnd enters l the compressive region, the applied stress intensity is too high i for the compressive stresses to stop, or even retard, crack growth.

Stress improvement of smaller diameter pipira will.not make .f conditions for unacceptable crack growth worse., For example, SI- j of 8-inch diameter piping with a 104.TW crack will essentially- 'j prevent further' crack growth, whereas a 10% TW crack in an '

as-welded joint will grow to an unacceptable depth within one operating cycle. After siaof'an 8-inch diameter joint, a crack

~

must be at least 72% TW before it can grow. .These comparisons-were based on calculations:using a 10.Kai operating stress.

Similar results were determined for: 10-inch diameter; pipe.-

~

The major conclusion of this' evaluation is that if a: crack will- ,

not grow to an unacceptable depth within an operating cycle in the l as-welded condition, the same crack;would not grow to an un-acceptable depth within an operating cycle if stress improved.

We are unaware of any instances of crack initiation or unaccept-able crack growth in a properly' stress improved weldment. We are aware of several instances of, supposed new cracks in SI'd welds ,

that were inspected in: successive outages. In the first outage, the indications were not dispositioned as.IGSCC; in the following outage,. indications were dispositioned as IGSCC. However, reviews of these cases concluded that the-interpretation of the first outage's data was incorrectLand'should have been dispositioned as ICSCC. We are also not aware of any cases of IGSCC being detected in a stress improved weldment-that contained no' reportable /

recordable indications in the previous outage inspection.;

1 '

l We consider that this evaluation shows that it is prudent to.

stress improve smaller diameter _. piping regardless of-whether or not 100% immediate post-SI inspection is performed. sI will improve the condition of-joints with no or shallow cracks and

'l l

will not-worsen the condition'of: joints;with-deeper cracks. We .

l. do not consider it technically prudent to avoid SI due to'the 1 1013-0012.19 W

1

,, TR .050 l Rev. 2 l

. Page 19 of 43 ALARA penalty taken to perform 100% immediate post-SI inspection ,

and the associated required weld crown reduction to facilitate i the inspection. Therefore, the GPUN Program does not require  !

100% post-sI inspection for welds <12 inches,in diameter (core spray and ICs excluding safe-ends). However, for welds which-will be both SI and inspected during the same refueling outage,.

the inspection will be performed, in a majority of cases, after  ;

SI. 4 4.3 Insoection of Cast Material GPUN does not consider that IGSCC in cast-stainless steel is a q l generic problem because there.are no reported cases of through-  :

j wall IGSCC in this material. The UT techniques developed to date s can only detect cracks with size. larger than 50% through-wall thickness in centrifugal cast stainless.stee1~(CC85). The grain .;

structure in the CCss is much more uniform than that found in  !

} statical cast stainless steel. It is the lack of uniformity that affects the ability to perform a reliable examination for small cracks such as IGSCC.. Therefore, we consider that there is no NDE j technique available that would reliably detect IGSCC and meet the requirements of GL 88-01 and NUREG 0313,-Rev. 2-[19).

! EPRI is working on developing screening methods to determine j acoustic characteristics of castings and the resultant methods j for performing reliable examinations.- GPUN will develop = plans to inspect casting welds when acceptable techniques are:available.

Since no meaningful'results would be generated by the' current NDE.

methods, ultrasonic inspection of these casting welds during 13R i would not be in the interests of reducing radiation exposure.

i l

By the and of the 14R outage,'only the 5 Recirculation-pump, suction ,

elbow-to-pump casing (all are castings) welds will not have been

stress improved. 'That is because heavy-walled piping should not be stress improved unless it can be adequately examined afterwards.

However, these 5-welds will;be protected by HWC. For these ,

reasons, visual inspection during pressure testing for these

^

5 Recirculation System welds will be performed >1n-lieu-of UT inspection. .

l

[

4.4 Imolementation of Hydrocen Water Chemistry j l GPUN has implemented Hydrogen Water. chemistry (HWC) 'during Cycl's 12 l (1990). Where NWC is effective in' flowing systems (Recirculation. ,

system and Reactor' Water Clean-Up System), a' factor of two reduc-tion in inspection frequency will be applied upon approval by~tte NRC.

l i

s t

013-0012.20 .li L . - .

. . . _ . - _ _ _ = - .. __ _. . ._

0 TR - 050, R3v. 2 Page 20 of 43 i l

Portions of systems that aria stagnant during power operation will also be afforded additional protection by HWC.- The existence of thermal mixing would provids this benefit.- systems considered to benefit in this way include shutdown cooling and-ICS condensate saturn lines, all of which are' attached'to the recirc. piping. For the GPUN Program, a factor of 2 reduction in inspection frequency will be applied in these cases upon approval by the NRC. '

l 4.5 RWCU Welds Outboard of the second Isolation Valve We have determined from a review of-the piping stress report, piping fabrication and intatallation records, and past inspection history that (a) the operational, piping stresses are predominately higher, approximately by a factor of 2, inboard of the second isolation valve. Therefore, welds residing inboard have.  ;

a higher propensity for IGSCC than those outboard of-the second valve; (b) there is no documentation that weld repairs exist l within the welds outboard of;the second valve which could increase the IGsCC propensity over those welds residing inboard of the ,

second valve. Additionally,.the welds located outboard of the .;

second isolation valve received similar RT during construction as. (

that of inboard welds; (c) tnere is no evidence of significant J

. piping material chemistry difference between the piping inboard and outboard of the second isolation valve. Thereford, the propensity for IGSCC in regard to alloy rcepocition is relatively the same; (d) the recults of our previous mucmented c: aminations conducted during the llR and 12R outages show no evicence of IGSCC of RWCU ,

welds inboard of the mecond valvo, -l Based on the above discussior . the resulta- of the previous augmented examination woald bound the conditions of the RWCU welds located outbosed of the seco>td isolation valve.= Therefore, we believe-it is prudent to implement a visual inspection program for welds located outboard of the second. isolation valve during the hydro test. However, in response;to the NRC Staff's generic concern foJ kWCU welds outboard of the second isolation valve, GPUN will inspect 10% of these welds in the 13R (19). Should; indications of IGsCc be found.in these welds, a required sample- 1 expansion as discuenednin paragraph.5.3.2 would be implemented.

l l

l

'l t 5

?.

l l

l l

1 013-0012.20 u

... TR - 050-Re'n 2 I Pa,e 21 of 43 l I

5.0 l[1DLEO 0313. REY. 2 5.1 NUREG 0313 ScoDe For Ovater Creek The scope of the'NUREG applies to'the following systems:

Svutem Extent No of Welds Recirculation Whole 89 Isolation Condenser Whole 189 Shutdown Cooling Stainless Steel Portion 14 Core Spray. Stainless Steel Portion 27 greater than 200'F-Reactor Water Clean-Up From Recirc. to inlet 136 side of non-regenerative heat' exchanger and from-outlet side of third re- .

generative heat exchanger shell~to Recirc.

Closure Head Three nozzles to points 7 at which diameter is less than 4" or changes to carbon steel ~.-

Table 1 lists the approximate number'of.wel'de included in the

( scope of the NUREG for each system both beforeL andLaf ter 13R.

In summary, 462 welds (26 of them are_not'inspectable) fall within the scope of the NUREG. Among the 26 uninspectable welds,-19 welds will be eliminated.through pipe spools. replacement.during ,

j 13R. Therefore, only 7 welds will not.be inspectable.after 13R.

The number of welds in each system by category, as well'as the.

number of inspections required for 13R4byfcategory,=are listed in Tables 2 thru 14. The reason for identifying; welds inside/outside y the drywell for the Isolation Condenser, System and the Reactor l Water Clean-up System is that the drywell is a-difficultiarea to

! schedule work'since it is;normally the most congested area during  ;

an outage.

5.2 GPUN ProDosed Procram >

Tables 2, 3, 4, 5, 6, 7, 8, 9, 10,,11, 12, 13,.14 end'15 describe ci the categories and inspection of. welds before, during and after  !

the 13R refueling outage. Tables 8, 9, 10 and 12' indicate a- i dramatic reduction in lower category. welds because of the replacement of the ICS piping,.the' replacement of the RWCU

! penetration piping, and the-cladding / replacement of the closure head welds. ~The above. tables also indicate the movement of many >.!

welds into category "C" due to stress improvement.

013-0012.21 ,

i f

l1

~

, -G, .-- ~: - -

. TR - 050 Rev. 2' Page.22 of 43 5.3 Samole Exoansion GL88-01 requires sample expansion if indications of IGScc are de-tected in the initial sample. The additional sample size should be approximately equal to that of the initial sample of the cate-gory of weld in which IGScc is detected, irrespective of sample  ;

and pipe size. If.IGSCC is detected in_the second sample, all l The sample expansien welds in that category should be inspected.

requirements of GL 88-01 will be partially met as modified-herein:

5.3.1 System Safe-Ends It is proposed that should flaws be detected in safe-end welds of a specific category, as defined in GL88-01, then an equal number of safe-end welds be examined within that category in the expansion sample. Should flaws be

-detected in this expansion sample, all safe-end welds in '

that specific category will be examined.'

5.3.2 RWCU PiDina It is proposed that 10% of the welds outboard of the second valves be included in the 13R UT inspection-sample. If indications of IGScc are found in these welds, GPUN will approach the NRC Staff on their disposition and any plans for expansion. Also, we tentativelyfplan to= ,

inspect a further 10% of RWCU welds =during future refueling outages, pending results of 13R inspections and-subject to possible mospection reduction as may be allowed by the use of HWc.

5.3.3 Remainina Welds (Recirculation. Core Sorav, Shutdown '

'Coolina, Isolation Condenser Inside and outside Drvwell, and closure Head Picino)

It is proposed that should' flaws be'detectedLin welds of a specific category, as defined in CL88-01, then an equal number of welds be examined within that category;in the  ;

expansion sample. Should flaws be detected;in'this; '

expansion sample, all welds.in thatLspecific; category will be examined.

• i 5.3.4 Samole Exnansion for "Susoect" Welds -

" Suspect" welds are those that required ' extensive ,; '

examination in order to disposition welds <as not containing. indications of IGSCC. . Fo'r ' example, in'llR) there were-eight recirculation system welds that required additional examinations that. led to the conclusion,/at:

that time, that the indications were' characteristic of root condition,' geometry, and/or -counterbore._ ' Experience ' i at other plants showed that these conditions often 1 i

.013-0012.22 i

7

• i

,, TR - 0$01  !

Rev. 2' Page 23 of 43-resulted in subsequent examinations concluding that the indications were really characteristic of IGscC. .,

~

Therefore, our initial sample of six welds in 12R was-selected from the eight " suspect" welds f rom ~11R. -

Two of the six were dispositioned in 12R as containing l IGSCC. There were no substantial changes in. location oro i signal characteristics from the 11R<to 12R examinations.

Since our sample expansion required examination of additional six welds, the two- remaining 11R " suspect" .i welds plus four othere were examined.7 one of the two: ,

remaining 11R " suspect" welds was determined to contain q IGSCC. This finding resulted in examination of the remaining 49 Category C Recirculation system' welds. 'No additional indications of IGSCC were detected.

Based upon the above sequence, we, consider thatLour approach to weld selection was based.on sound engineering-. '

and EPRI-qualified NDE examiners'1 judgement. 'Therefore,:

we consider that the following sample expansion requirement for " suspect" welds to be technically sound ,

and ALARA conscious  ;

During each outage, a separate group of " suspect" welds, if there is any, will be established. Sample expansion of-these." suspect" welds.would not be required since all-the

" suspect" welds will be reinspected-for IGSCC during tho' following outages until:they can be determined.not i to have indications of IGSCC. ,In the event _that the " suspect" welds are determined to.be free ~of'IGSCC,*they will )

reenter the normal sampling' plan. 0 1

i Currently, there are no'" suspect" welds that require; i reinspection during 13R. ,

l 6.0 M'EER GENERIC 1.EXTER 88-01 RESPONSES REQUIRED L

6.1 The NRC Staff has proposed a change to the Technical Specifica-tions.(TS) to include a statement'in the section onlIs! that the.

'In-service Inspection Program-for piping covered by the scope of NUREG 0313 will-be in conformance with the NRC' positions on ..

l schedule, methods and personnel, Land' sample expansion included in ,

GL-88-01. The staff has also recognised that the:In-service Inspection =and Testing sections may be removed ~from the'TsLin the I l i

, d I

013-0012.24 '

l'

'* , f

'_-______. .a. , s- . . _ , _ . _ . ., , . ., ,

J!

. l

,. TR - 050 i Rev. 2 1 l

Page 24 of 43 I i:

l future in line with the Technical. Specifications Improvement. 1 Program.- If this does come into existence, then this requirement j

would remain with the ISI section when it is included in an a alternative document.

K To comply with the requirements of GL88-01, GPUN will submit a TS Change Request consistent with. this response during the upcoming refueling outage, scheduled for completion in the-Spring of 1991. }

6.2 The NRC Staff position in GL 88-01 is that leakage detection-

! systems should be in conformance with position C of Regulatory .

i Guide 1.45 " Reactor Coolant Pressure Boundary Leakage Detection I systems" or as otherwise previously. approved by the NRC.

Leakage detection systems for oyster Creek were reviewed by the NRC Staff during the Systematic Evaluation Program and-the results were documented in Section 4.16.2 of' Integrated Plant '

Safety Assessment Report -for oyster . Creek, NUREG-0822 dated January, 1983. The actions identified in that' report.have been-completed with the exception of the' airborne: particulate and }i gaseous radiation monitoring. system-(APCRMS). GPUN's recent submittal of July 1, 1988, states that installation of a new APCRMS will be completed during the operating cycle 12. The-submittal also identifies that-there are several leak detection-methods available for unidentified _ leakage into the containment sump at oyster Creek which operate on diverse principale.

l The normal method of monitoring unidentified' leak rate is to obtain flow integrator readings from the: containment sump' pump discharge every four' hour period and' calculate average flow rate. Approximately 1 gpm can be measured in a four hour interval. This methodology-is identified in oyster Creek l-Technical Specifications as the primary method-of leakage measurement. :I When the flow integrator is not available,'the average leakage 'I rate can be calculated using the known volume between the high and the low level alarms for the sump and the time required to fill the sump between these levels.

A recorder available in the control room also providea con-tinuous indication of an estimated unidentified leak rate to the containment sump by utiliaing a differential pressure r signal as a result of the sump level change. The sensitivity 1 l of the recorder is approximately 0.2 gpm.

Additionally, a timer available in the 480 volt switch gear '

room,provides the:run' time of:the containment sump pumps. .

This run time along with'the estimated flow rate of the sump-pumps can provide approximate leak rates. This methodology;is

, utilized every four hours during power operation.

i

,013-0012.25 '

i

,,- TR - 050~

Rev. 2 Page 25 of 43 Also, an annunciator will alarm in the control room if the time to fill the containment-sump is too short an interval. t The time associated with this alarm is set to bring in the-alarm if unidentified leak rate' equals or exceeds 4 gpm.

These methods provide quantitative indications of unidentified RCS leakage inside containment and also' provide assurance that un-identified leakage can be detected and quantified during cycle 12 operation pending operability of the new APGRMS. -

l The NRC Staff position was further amplified in GL 88-01 by-

• additional criteria as follows:

1. Plant shutdown should be initiated for inspection'and cor-rective action when,;within<any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less, . ,

any leakage detection system indicates an increase in. rate of .

unidentified leakage in excess of 2 gpm or its equivalent, or when the total unidentified leakage attains a-rate of 5 gpm or' equivalent, whichever occurs.first. For sump level monitoring ,

systems with fixed-measurement-interval methods, the level' should be monitored at approximately 4-hour intervalo or less.- l

2. Unidentified leakage should include.all leakage other thans a) leakage into closed systems, such as pump seal or valve packing leaks that are captured, flow metered, and I conducted to a sump'or collection tank,'or j

b) leakage'into the containment atmosphere from sources;that  ;

are both specifically located and known either not'to.

interfere with'the operations of unidentified leakage monitoring systems or not to'be from a'through-wall crack 'l in the piping within the reactor coolant pressure boundary.

3. For plants operating with an IGSCC Category D, E, F, or G welds, at-least one of the leakage-measurement' instruments associated with each sump shall be operable, and the outage time for inoperable instruments shall be limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or immediately initiate an orderly l shutdown. p By Amendment 97 to Provisional operating License No. DPR-16 for Oyster Creek, the limiting conditions for operation and surveillance requirements were authorized for the Reactor .

Coolant System leakage. This amendment added two new definitions (identified and unidentified leakage).to TS Section 1.0; revised TS 3.3.D to include.LCo's for;the containment sump flow monitoring system and the equipment; drain tank monitoring system; and added a new surveillance section TS 4.'3.H. This amendment incorporated GPUN's response-dated September 8, 1983, to IE Bulletin 82-03.

013-0012.26

}

• • TR - 050 Rev.12; l

. Page 26 of 43

]

I O 1 On March 17,-1987, GPUN submitted Technical-Specification Change Request #158 which adds additional conservatism to these requirements by proposing to;1imit the unidentified _ .

leakage for the Reactor Coolant System to a maximum leak rate increase of 2 gpm within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.while operating at steady state. power.: On May 23, 1989, the~NRC staff approved ,

this request as Amendment 133.- This amendment addresses <

item.1-of the NRC Staff position.

6.3 GPUN plans to notify the NRC of any flaws identified that do not meet IWB-3500 criteria of Section.XI of the code for continued operation without evaluation, or a change found in the condition of the welds previously known to be cracked, and our evaluation of-the flaws for continued operation and/or repair plans., GPUN will' l obtain NRC approval of the evaluations and/or repairs prior.to restart.

(

7.0

SUMMARY

In response to the NRC-staff's concern that all Category G welds, as a minimum, .tus inspected no later than the end of 13R, GPUN has revised its ,

inspection plan. Even though GPUN is not in complete compliance with '

NUREG 0313, this revised plan is-sensitive to the NRC concern while maintaining a controllable outage work scope and minimizing radiation exposure. Except for the recirculation inlet nozzles'and the RWCU piping outside of the second isolation valve, the proposed plan'will-result in all Category G welds being inspected by the end of-13R outage.

8.0 REFERENCES

1. USNRC Generic Letter 84-11, " Inspections of BWR Stainless Steel Piping," April 19, 1984.

l 2. NUREG-0313 Rev. 2, " Technical Report on Material Selection and l Processing Guidelines for BWR Coolant Pressure Boundary Piping,"  ;

USNRC, January 1988.

t

3. USNRC Generic Letter 88-01,."NRC. Position on4 IGSCC in BWR Austenitic Stainless Steel Piping," January 25,21988. .

4. USNRC IEB No. 82-03 Rev. 1, " Stress Corrosion Cracking in l Thick-Wall, Large-Diameter, Stainless Steel,' Recirculation System '

Piping at BWR. Plants," October 28, 1982. t

5. GPUN Topical Report No. 012 Rev. .1, " Oyster Creek Recirculation- 'I System' Piping Inspection Program," September 5L 1983.

l 4

013-0012.26 1

TR - 050 ,

.. R;v. 2'

-Page 27 of 43

6. GPUN Technical Data Report No. 580 Rev. 2, " Isolation Condenser System Piping Cracked Welds-Repa1 and Failure Analysis," 11-5-85. '

7. GPUN Topical Report No. 039 Rev. O, " Oyster Creek Cycle 11R Outage IGSCC Activities," 9-30-86.

8. Oyster Creek FSAR Amendments 29, 35, 36, 37, 40, 43 and 47.

i

9. NRC Letter dated October 18, 1988.

10. NRC/GPUN Telecon - 10/25/88.

11. NRC/GPUN Telecon - 11/30/88.

12. GPUN GL88-31 Response ~- 8/12/88.

13. GPUN GL88-01-Response 1/31/89.

14.. GPUN GL88-01 Response 11/16/89.

15. General Electric Report No. 89-178-002 " Evaluation of Core Sample NG-D-11 from Oyster Creek Nuclear Generating Station."

16. General Electric Report No. 89-178-007 " Meta 11uro!:al Examination' of 2-inch Head Cooling Line Nossle Weld from Oyster. Creek-Generating Station."

17. GPUN Letter to NRC No. 5000-90-1891 " Oyster Creek-Nuclear-Generating Station Docket.No.-50-219, IGSCC Inspection Plan -

RWCU" dated 2/21/90.

18. NRC Letter' dated April 17, 1990.

19. GPUN Response Letter to NRC No.- 5000-90-1938 dated 6/7/90.

9.0 FIGURES

1. Safe-end configuration.

2. Comparison of crack growth rates.

3. Peach Bottom Unit 3. '

t

4. Effects of aqueous impurities on crack growth. 3 i

s.

J 013-0012.27 c

l

=

'TR - 050  ;

Rov. 2 '

Page 28 of-43 i t

i e

r t

w eiestwo euntil a wie SilcashMt (DT intl l4 t/4 e t 1ie . e,) . . '-

l e.ese samem -

IE e.m noe tassans: 4

- Wiles tePt se e,dhill s' ass me s*w,usse 61 stumaana easts8.37%

i.aes samme MIIAE' .

e.same testasma si .esem w as .

,,,,,, ,,,,,,g IIsmas lateA -

M '[w Y ' IIrEss*Al 7

- rM.*eu,[,."".

/e/ ' y / q{;q -

/ucui cau sam i / a'*i-

. i,,,,,

x _'

i 'itumais e

p f / 1/ ' u in reis asun simasu seassas- 88,

\/

i H/ Am/'a a4*3

/- /( unia/i._f ,

M \ =. un nei m I l\ \ \ ,n X g

.A N \\

usm ,,zq l l

h. l\ \. ,

q i r

' /l

,, ,sisiatu et is[, '

,,,,Y

-Ilw an i y .

}.

l  :,, -- .-- ..~.

l

....,.,u,,.,.

.e x i. i.ni u. . ,

- u. I. I wu nas a

,3 g yg , ,to i soie. .I' -1 ma am.se 1 4.e. t w g f wt us (mies reica io l eim secusa.

.\

i

! l l <

1 1

j, FIGURE 1 - RECIRCULATION SAFE-END CONFIGURATION l

013-0012.28 l-

TR - 0$0 Rev. 2

'Page 29 of 43 i

t l

l i

l 1

,i 28 F '}

30 -

Ag cond -0 2 me/cm Normei SWR 1s - weser cheneery $

• 24 N

'e peach noteo 33 p j

-.g to - /

, Ag cond -0.3,e/cm e ..

I s -

r Ane cond -0.s

/

2 * * **'Nr

% water chemetry-

. vo,eede 3,

1. -e8. . gg N Ae9 cond -0,I me/cm -

de-0^ 54 l

• 8" icos isas sono I 2soo Test Time (hrst ,

% M Comparison og she creek erowe reeen W

• • ""*** type aos eenwone essee ,eem,ne m 7""'8'*esa we8er chamewy as oreemen.2.d *e8er themetry t

FIGURE 2 013-0012.29 i

-____----3 TR'- OSO Rev. 2 Page 30 of 43 1

i o

l t

l PEACM DOfftlHA UNIT 3 Soneesed 1)pe ate SS Crack Langth On4 Ceneuctivey (us/cen)

• IS4 2.3 e

i

.753 t 4

A. e 3.3 3.37 a 1(P4 inA =

2.0 .

b

.783 =

M' O e em a 104inA = 1.5

.751 -

M

• M = 1.0 inWueiOR W4flal0A No,1 No. 2 1

.rio -

- o.5 1

l -

L. U f L T

i

.re 1 I i 1 o-i tion mas men seco- . anos anos tres asco TINIS N

• i i

t I. '

i i

i FIGURE'3 1.

l i.

t i

013-0012.30' t

i s E ,

J'-...

..*= ,

TR - 050-  !

Rov. 2 -

_ Page 31 of'43 .

i i

cyclic

/ 4040

% ComefAmf LOAS

. . a,, *  %

.. <. - .-  % < . . . . . - 0, .

f l -

.31 -

3

// .' '

.s2 - n .

g I

w '3

--,- - .- i

- - - assess onvoam - -- - -- - .Enessouvesa. .

2 3-7.1 w -

=>

g -

/- =

sessononvosas - . V'

... [ -

s

"' .s

.79 * -

-t

_a -

30

.?8 -

,p7 ,

1_ . 4 .

La

. . . .a 4 A

4. -

.1 ' ' ,

See

-A-Inee tSee. Pese .

110A IId ' '

FIGURE 4.

EFFECT OF AQUEOUS IMPURITIES ON CRACK GROWTH OF SENSITIZE TYPE 304 STAINLESS STEEL IN 550' (288'C) WATER 013-0012.31 5

+ ~ ,, .m - ,

1

• I TR - 050 4 R3v. 2 i Page 32 of 431

{

i 1

10.0 IA3Lg3

1. Number of Welds in Scope of NUREG 0313. Rev. 2

2. Recirculation Safe-End Welds

3. Core Spray safe-End Welds f

4. Isolation condenser Safe-End Welds-  !

5. Recirculation Pipe Wolds

6. Shutdown Cooling Pipe Welds

7. Core Spray Pipe Wolds

8. Reactor Water Clean-up Inside Second Isolation Valve

9. Isolation Condenser Pipe Welds (Inside Second Isolation Valve)

10. Closure Head Pipe Welds. '

11. Reactor Water Clean-up (Outside Second Isolation Valve)' t

12. Isolation conder.ser Pipe Welds (Outside Second Isolation Valve) _i

13. Safe-End Walds

14. Summary of Total Inspectable Pipe Welds '

15. Inspection Schedules for BWR. Piping Weldments Kev to Inspection cateoories A. Resistant Materials j B. Nonreslutant' Material SI within 2 years of operation >

C. Nonresistant materials SI after 2 years'of operation and Post <

Process-Inspected. +

D. Nonresistant material - Inspected - No SI.

. , :3 C/D.. Nonresistant materials SI.after 2 years'of operation and not Post -

Process Inspected, but has been inspected for TGSCC during'11R or.  !

12R. .This category will be< treated as' Category."D". -i C/G. Nonresistant materials SI after-2 years of operation and-not Post-

' Process Inspected and not inspected for IGSCC during 11R or 12R.

This category will be treated as Category "G".

E. Cracked - Overlayed or SI l

[

j F. Cracked - Inadequate - No Repair i

• G. Nonresistant - not inspected l

l r

013-0012.32

9

.. TR - 050 s,* R;v. 2 Page 33 of 43 TAntt 1 NUMBER OF WELDS IN SCOPE OF NUREG-0313. REV. 2 '

BEFORE AND AFTER 13R I

INSPECTAkt2 WELDS UNIN- TOTAL IN-TOTAL INSIDE OUTSIDE SPECTABLE SPECTABLE INSIDE SYSTEM OUTSIDE 2ND~ ISO. -2ND ISO.

WELDS WELDS WELDS DRYWELL - DRYWELL- VALVE VALVE-B A B A 'B A B A B A-~ B A B A Recire. 89 89 *5 *5 84 84 84 84 0 [0 '84 84 0 0 RWCU 136 131 ** 5 0 131, 131- 30 30 101 101 46 '46 85 85 I CS 27 27 0 0 27 27 27 27 0 0 '27 27 0 0 SDC 14 14 0 0 14 14- 14 14 0 0 *i 14 0 0 IC 189 114 16 2 173 112 44 44 -129 68 '44 48

• • • 129 64 ,

Closure Head 7 7 0 0 7 7 7 7 'O O 7 7 0 0 TOTAL 462 382 26 7 436 375 206 206 .230 169 '222 226 214 149 l

casting-to-casting welds.

Welds inside penetrations (will be eliminated in 13R). '

i i

i 8 welds inside penetrations, 4 flued head-to-valve welds, 2 casting-to-casting-welds, 2' saddle welds (ICS piping outside drywell will be replaced during 13R). I A After 13R ,

B: Before-13R' 1 i

i 013-0012.33 0

rj

1

]

TR - 050-4,* R;v. 2 I Page 34 of 43 l l ,

i- I l- .;

TABLE 2 f

l RECIRCULATION '

SAFE-END WELDS -

i l l <

NUREO TOTAL WELDS TOTAL WELDS 3 INSPECTION PRIOR TO 13R. AFTER  !

CATEGORIES '13R INSPECTIONS 13R C-12R/HF 4 0- 4 C-13R/HF 0 0 -8 .

C/HF 16 8 '8 I TOTAL 20 8 20 '

HF - HWC Flowing i 4

TABLE 3 ,

( CORE SPRAY SAFE-END WELDS NUREG TOTAL WELDS TOTAL' WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-12R/NP 6 0 6 TOTAL 6- 0 -6 i

NP - No HWC Protection. i l

l .L i

i t

-i

i 013-0012.34

, i 4,

e TR - 050 i

• • R v. 2 Page 35 of 43 )

l

)

TABLE 4 ISOLATION CONDENSER SATE-END WELDS NUREG TOTAL WELDS TOTAL WELDS

' INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-12R/NP 4 0 4 TOTAL 4 0 4 f

NP- No HWC Protection 5 i

TABLE 5 RECIRCULATION PIPE WELDS NUREQ TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-11R/HF 58 0 58 E/HF 6 3 6

TOTAL 64 3 64 HF - HWC Flowing From Vessel Outlet to Vessel Inlet

^

i 013-0012.35 t

l'

e i I

F TR - 0$0-s' R3v. 2 I Page 36 of 43 i

.  ?

I 2

i i TABLE 6 SHUTDOWN COOLING PIPE WELDS  :

-)

l NURtc TOTAL WELDS I INSPECTION . TOTAL WELDS PRIOR TO 13R

}

CATECORIES AFTER '

13R INSPECTIONS

?

13R  !

C-13R/NP 0. 0 $ h C-13R/Hs 0 0, 9 '

D/NP 4  ;

4- 0 >

D/MS 5 $ 0 -

C/NP 1

1. 0 'l 0/HS 4 4' 0 I TOTAL 14 r' 14 14 5 HS - HWC stagnant with protection; based,on thermal mLxing of the &

recirculabon connection point.

NP - No NWC Protection I

l, i

.rt f

l

-l "

i i

i s

4 h 013-0012.36-

\

?

a,

-E

TR - 050 '

s' Rev. [ I Page 3? of 43 l t

TAELE 7 l CORE SPRAY l PIPE WELDS .

I NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER f CATEGORIES 13R INSPECTIONS 13R ,

i C-12R/NP 8 0 14 C-13R/NP 0 0 2 '

C/D-12R/NP 7 3 4 C/G/NP 3 3 0 D/NP 2 2 0 E/NP 1 0 1 i

TOTAL 21 8 21 NP - No HWC Protection TARLE 8 REACTOR WATER CLEAN-UP PIPE WELDS (INSIDE SECOND ISOLATION VALVE)

NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIEE 13R INSPECTIONS 13R A 0 0 8 l D/HF 20 9 38 G/HF 26 ' 19 0 TOTAL 46 28 46 l

l HF - HWC Flowing from recirculation supply to return connection points Note All Category "A" welds installed in 13R will be baseline inspected j (S welds) 013-0012.37

- - . . . _ ..y -y9 9

_. . _ _ . . _ . . . _ . _ . . ~ . . _ _ _ _ _ __

e l

.,, TR - 050' >

RN. 2 I Page 38 of 43 l

• i TABLE 9 ISOLATION CONDP et.i [

t PIPE WELLS *

(INSIDE SECOND ISOLAT10N VA?.VE) I NUREG TOTAL WELDS 1

TOTAL WELDS

• INSPECTION PRIOR TO 13R AFTER

• CATEGORIEB 13R INSPECTIONS 13R A/Hs 0 0 12 ..

-j C-12R/NP O O 8 I

C-13R/HS 0 0 14

• f C/D-12R/NP 5 4 1 C/D-13R/HS 0 0 9 D/HS 15 4 0' '

(

D/NP 2 0 0 G/HS 12 10 0 '

G/NP 2 0 0 '

C/0/NP 4 4 .O TOTAL 40 22 44 i I

r NP - No NWC Protection ,

'f HS - HWC stagnant with protection based on' thermal' mixing at -

the recirculation connection pointL_ r Note i All Category "A" welds installed in:13R will be baseline inspected (12 welds)- '!

, t i

f

.i,r l

ll

.?

i 013-0012.38 lh i '

+4 TR - 0$0 O Rev. 2 i Page 39 of 43 TABLE 10 CLOSURE HEAD PIPE WELDS NUREQ TOTAL WELDS

TOTAL WELDS INSPECTION PRIOR TO 13R CATEGORIES AFTER 13R INSPECTIONS 13R A 0 0 7 D/NP 4 0 0 O/NP 3 0 0 t

TOTAL 7 0 7 NP - No HWC Protection Note All Category "A" inspected (7 welds) welde installed in 13R will be baseline TABLE 11 REACTOR WATER CLEAh-UP PIPE WELDS (OUTSIDE SECOND ISOLATION VALVI)

NUREO TOTAL WELDS INSPECTION TOTAL WELDS PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R D/HF 0 0 9 O/HF 85 9 76 TOTAL 85 9 85 l

HF -

HWC return Flowing from recirculation connection pointe supply to 013-0012.39

~ )

, e .

,e TR - 050 0 R;v. 2 Page 40 of 43 i

TABLE 12 ISOLATION CONDENSER PIPE WELDS l' (outsIDE sECoND ISOLATION VALVE) '

i NUREG TOTAL WELDS TOTAL WELDS ,

INSPECTION PRIOR TO 13R AFTER CAI[GORIE2 13R INSPECTIONS 13R A/NP O O 64 D/NP $9 0 0 E/NP 22 0 0 G/NP 48 0 0 TOTAL 129 0 64 NP - No HNC Protection Note: All Category "A" welds installed in 13R will be baseline inspected (64 welds)

{

TABLE 13 SAFE-END WELDS NUREG TOTAL WELDS TOTAL WELDS INSPECTION PRIOR TO 13R AFTER CATEGORIES 13R INSPECTIONS 13R C-12R/HF 4 0 4 C-12R/NP 10 0 10

• C-13R/HF 0 0 8 G/HF 16 8 8 TOTAL 30 '

8 30 HF - HWC Flowing , ,

NP - No HWC Protection l

j 013-0012.40 f

,i

. e i

.* TR - 050

1. R w. 3 l e

Tage 41 of 43 i 1

TABLE 14 1 GPUN PROGRAM l

SUMMARY

OF TOTAL INSPECTABLE WELDS j

. i NUREG TOTAL WELDS TOTAL WELDS i INSPECTION PRICR TO 13R AFTER l CATEGORIES 13R INAPEcTIONS 13R j i

A 0 0 91

)

C-11R/HF 58 0 58 C-12R/HF 4 0 4 C-12R/NP 18 0 32 C-13R/HS 0 0 23 C-13R/HF 0 0 8 C-13R/NP 0 0 7 C/D-12R/NP 12 7 5 C/D-13R/HS 0 0 9 D/HF 20 9 47 D/HS 20 9 0 D/NP 71 6 0 E/HF 6 3 6 E/NP 23 0 1 0/HF 127 36 84 G/HS 16 14 0

! O/NP 54 1 0 C/0/NP 7 7 0 TOTAL 436 92 375 HF - HWC Flowing HS - HWC Stagnant - with protection via thermal mixing NP - No HWC Protection -

Note All Category "A" welds installed in 13 R will be baseline inspected-(91 welds).

013-0012.41

• • TR - 050

%' Rev. 2 Page 42 of 43 l .

I i

Tant e 15

(

INEPECTION SCHEDULES FOR RWR PIPING Wet.nMEk73 (3}

DESCRIPTION IGSCC OF WELDMENTS INSPECTION GPUN PROPOSED NOTER CATEGORY EXTENT /REMEhULE EXTENT / SCHEDULE Resistant Materials A 25% overy 10 years i SAME (at least 12% in 6 years) ,

Nonresistant Matla. (1) B 50% overy 10 years 1

SI within 2 years SAME of operation (at least 25% in 6 years) 1 i i

Nonresistant Matle. (1) C All next 2 re-SI after 2 years fueling cycles, then SAME 5 of Operation All every 10 yrs '

l (at least 50% in 6 yrs) j Nonresistant Matl. (1) D All every 2 refuel-No SI SAME ing cycles ,

Cracked (1) E 50% next outage, then overlayed or SAME SI All every 2 refuel- ,

ing cycles l Cracked F Inadequate All overy refueling SAME No Repair outage i

Nonresistant (2) o Not Inspected All next refueling All by and of 13R outage outage except 8 recirculation system safe-ends welds and RWCU welds located outboard of the second isolation valve (See para. 2.0).

013-0012.42

o i o t s e* TR - 050 k' R v. 2' Page 43 of 43 i 4

6 i

TABLE 15 ( Cont ._ ) '

Notess t (1)

All welds of non-resistant material should be inspected after a 6 stress improvement process as part of the process.  !

should be followed after this initial inspection. Schedules shown I (2)  !

Welds that are not UT inspectable should be replaced,-" sleeved," or local leak detection applied. j for leaks may also be considered. RT examination or visual examination ,1 I

i r

i i

i E

I f

i i

~i

p b

i

\

1

s i,

f i

013-0012.43'  !

. _. . _ ~ . , . . , , _ . . .- - . . ., .- ,